• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.1
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• pp.65-75
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• 2006

Transparency on the Total System Performance Assessment (TSPA) is the key issue to enhance the public acceptance for a radioactive repository. To approve it, all performances on TSPA through Quality Assurance is necessary. The integrated Cyber R&D Platform is developed by KAERI using the T2R3 principles applicable for five major steps : planning, research work, documentation, and internal & external audits in R&D's. The proposed system is implemented in the web-based system so that all participants in TSPA are able to access the system. It is composed of three sub-systems; FEAS (FEp to Assessment through Scenario development) showing systematic approach from the FEPs to Assessment methods flow chart, PAID (Performance Assessment Input Databases) being designed to easily search and review field data for TSPA and QA system containing the administrative system for QA on five key steps in R&D's in addition to approval and disapproval processes, corrective actions, and permanent record keeping. All information being recorded in QA system through T2R3 principles is integrated into Cyber R&D Platform so that every data in the system can be checked whenever necessary. Throughout the next phase R&D, Cyber R&D Platform will be connected with the assessment tool for TSPA so that it will be expected to search the whole information in one unified system.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.15 no.2
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• pp.151-159
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• 2017

A deep geological disposal at a depth of 500 m in stable host rock is considered to be the safest method with current technologies for disposal of spent fuels classified as high-level radioactive waste. The most important requirement is that the temperature of the bentonite buffer, which is a component of the engineered barrier, should not exceed $100^{\circ}C$. In Korea, the amount of spent fuel generated by nuclear power generation, which accounts for about 30% of the total electricity, is continuously increasing and accumulating. Accordingly, the area required to dispose of it is also increasing. In this study, various duplex disposal concepts were derived for the purpose of improving the disposal efficiency by reducing the disposal area. Based on these concepts, thermal analyses were carried out to confirm whether the critical disposal system requirements were met, and the thermal stability of the disposal system was evaluated by analyzing the results. The results showed that upward 75 m or downward 75 m apart from the reference disposal system location of 500 m depth would qualify for the double layered disposal concept. The results of this study can be applied to the establishment of spent fuel management policy and the design of practical commercial disposal system. Detailed analyses with data of a real disposal site are necessary.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.2
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• pp.133-141
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• 2020

A geological repository comprises a natural barrier and an engineered barrier system. Its design components consist of canisters, buffers, backfill, and near-field rock. Among the engineered barrier system components, bentonite buffers minimize the groundwater flow from near-field rock and prevent the release of nuclide. Investigation of the hydraulic conductivity of the buffer to groundwater flow is an important factor in the performance evaluation of the stability and integrity of the engineered barrier of the repository. In this study, saturated hydraulic conductivity tests were performed using Gyeongju bentonite at various dry densities and temperatures, and a hydraulic conductivity prediction model was developed through multiple regression analysis using the 120 result sets of hydraulic conductivity. The test results showed that the hydraulic conductivity tends to decrease as the dry density increases. In addition, the hydraulic conductivity increased with increasing temperature. The multiple regression analysis results showed that the coefficient of determination (R²) of the hydraulic conductivity prediction equation was as high as 0.93. The hydraulic conductivity prediction equation presented in this study could be used for the design of engineered barrier systems.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.3
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• pp.329-338
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• 2019

Suction of unsaturated bentonite buffers is a very important input parameter for hydro-mechanical performance assessment and design of an engineered barrier system. This study analyzed suction measurement techniques and constitutive models of unsaturated porous media reported in the literature, and suggested suction measurement techniques and constitutive models suitable for bentonite buffer in an HLW repository. The literature review showed the suction of bentonite buffer to be much higher than that of soil, as measured by total suction including matric suction and osmotic suction. The measurement methods (RH-Cell, RH-Cell/Sensor) using a relative humidity sensor were suitable for suction measurement of the bentonite buffer; the RH-Cell /Sensor method was more preferred in consideration of the temperature change due to radioactive decay heat and measurement time. Various water retention models of bentonite buffers have been proposed through experiments, but the van Genuchten model is mainly used as a constitutive model of hydro-mechanical performance assessment of unsaturated buffers. The water characteristic curve of bentonite buffers showed different tendencies according to bentonite type, dry density, temperature, salinity, sample state and hysteresis. Selection of water retention models and determination of model input parameters should consider the effects of these controlling factors so as to improve overall reliability.

• Tunnel and Underground Space
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• v.31 no.6
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• pp.561-577
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• 2021

The compacted bentonite buffer in a geological repository for high-level radioactive waste disposal is saturated due to groundwater inflow. Saturation of the bentonite buffer results in bentonite swelling and bentonite penetration into the rock discontinuities present around the disposal hole. The penetrated bentonite is exposed to groundwater flow and can be eroded out of the repository, resulting in bentonite mass loss which can affect the physical integrity of the engineered barrier system. Hence, the evaluation of buffer-rock interactions and coupled behavior due to groundwater inflow and bentonite penetration is necessary to ensure long-term disposal safety. In this study, the effects of the bentonite penetration and swelling on the physical properties of jointed rock mass were evaluated using the quasi-static resonant column test. Jointed rock specimens with bentonite penetration were manufactured using Gyeongju bentonite and hollow cylindrical granite rock discs obtained from the KAERI underground research tunnel. The effects of vertical stress and saturation were assessed using the P-wave and S-wave velocities for intact rock, jointed rock and jointed rock with bentonite penetration specimens. The joint normal and joint shear stiffnesses of each joint condition were inferred from the wave velocity results assuming an equivalent continuum. The joint normal and joint shear stiffnesses obtained from this study can be used as input factors for future numerical analysis on the performance evaluation of geological waste disposal considering rock discontinuities.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.10 no.3
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• pp.133-141
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• 2012

The impacts influenced on the performance and safety of a repository are classified as units of Features, Events, and Processes (FEP), for the total system performance assessment (TSPA) related to the permanent disposal of HLW. The importance is evaluated in consideration of the frequency, consequence, regulation, suitability of a specific site, etc. and then these are grouped as a similar FEP. A scenario describing the migration of radionuclide from the repository to the biosphere is derived from understanding the interaction among these groups. KAERI has developed the KAERI FEP lists by review and collation of the foreign studies. The KAERI FEP list has been reviewed by several Korean experts. The five major scenarios describing possible future evolutions of the geological disposal system have been developed by RES and PID methods. Also the CYPRUS which is a KAERI integrated database management system for the total system performance assessment (TSPA) related to the permanent disposal of HLW has been developed and the results of the FEP and scenario development have been uploaded in this system.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.7 no.2
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• pp.101-107
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• 2009

Investigation of resaturation and thermal-hydro-mechanical behavior for the buffer of a repository requires measuring the water content of compacted bentonite. This study investigated the relative humidity of compacted bentonites using a humidity sensor (Vaisala HMT 334) applicable under high temperature and pressure, and then conducted a multi-regression analysis based on the measured results to determine relationships among the water content, relative humidity, and temperature. The relationships for the compacted bentonites with the dry densities of 1,500 $kg/m^3$ and 1,600 $kg/m^3$ were expressed as ${\omega}=0.196RH-0.029T+1.391({r^2=0.96)}$ and ${\omega}=0.199RH-0.029T+2.596({r^2=0.98)}$, respectively. These were then used to interpret the resaturation of bentonite blocks in the KENTEX test.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.6 no.2
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• pp.147-154
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• 2008

It is expected that the temporary storage facilities at the nuclear power plants will be full of the spent fuels within 10 years. Provided that a centralized interim storage facility is constructed along the coast of the Korean peninsula to solve this problem, a substantial amount of spent fuels should be transported by sea or by land every year. In this paper we developed a computer program for the analysis of transportation logistics of the spent fuels from 4 different nuclear power plant sites to the hypothetical centralized interim storage facility and the final repository. Mass balance equations were used to analyze the logistics between the nuclear power plants and the interim storage facility. To this end a computer program, CASK, was developed by using the VISUAL BASIC language. The annual transportation rates of spent fuels from the four nuclear power plant sites were determined by using the CASK program. The parameter study with the program illustrated the easiness of logistics analysis. The program could be used for the cost analysis of the spent fuel transportation as well.

• Journal of the Korea Concrete Institute
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• v.25 no.3
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• pp.321-329
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• 2013

Recently, the medium-low level radioactive waste from nuclear power plant must be transported from temporary storage to the final repository. Medium-low level radioactive waste, which is composed mainly of the liquid ion exchange resin, has been consolidated with cementitious material in the plastic or iron container. Since cementitious material is brittle, it would generate cracks by impact load during transportation, signifying leakage of radioactive ray. In order to design the safety transporting equipment, there is a need to check the compressive strength of the current waste. However, because it is impossible to measure strength by direct method due to leakage of radioactive ray, we will estimate the strength indirectly by the dynamic modulus of elasticity. Therefore, it must be identified the relationship between of strength and dynamic modulus of elasticity. According to the waste acceptance criteria, the compressive strength of cement based solid is defined as more than 3.44 MPa (500 psi). Compressive strength of the present solid is likely to be significantly higher than this baseline because of continuous hydration of cement during long period. On this background, we have tried to produce the specimens of the 28 day's compressive strength of 3 to 30 MPa having the same material composition as the solid product for the medium-low level radioactive waste, and analyze the relationship between the strength and the dynamic modulus of elasticity. By controling the addition rates of AE agent, we made the mixture containing the ion exchange resin and showing the target compressive strength (3~30 MPa). The dynamic modulus of elasticity of this mixtures is 4.1~10.2 GPa, about 20 GPa lower in the equivalent compressive strength level than that of ordinary concrete, and increasing the discrepancy according to increase strength. The compressive strength and the dynamic modulus of elasticity show the liner relationship.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.5 no.3
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• pp.229-238
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• 2007

In this paper, structural design requirements and safety evaluation criteria of the spent nuclear fuel disposal canister are studied for deep geological deposition. Since the spent nuclear fuel disposal canister emits high temperature heats and much radiation, its careful treatment is required. For that, a long term(usually 10,000 years) safe repository for the spent nuclear fuel disposal canister should be secured. Usually this repository is expected to locate at a depth of 500m underground. The canister which is designed for the spent nuclear fuel disposal in a deep repository in the crystalline bedrock is a solid structure with cast iron insert, corrosion resistant overpack and lid and bottom, and entails an evenly distributed load of hydrostatic pressure from underground water and high pressure from swelling of bentonite buffer. Hence, the canister must be designed to withstand these high pressure loads. If the canister is not designed for all possible external loads combinations, structural defects such as plastic deformations, cracks, and buckling etc. may occur in the canister during depositing it in the deep repository. Therefore, various structural analyses must be performed to predict these structural problems like plastic deformations, cracks, and buckling. Structural safety evaluation criteria of the canister are studied and defined for the validity of the canister design prior to the structural analysis of the canister. And structural design requirements(variables) which affect the structural safety evaluation criteria should be discussed and defined clearly. Hence this paper presents the structural design requirements(variables) and safety evaluation criteria of the spent nuclear fuel disposal canister.